Transportation and shipment of spent nuclear fuel for eventual disposal is regulated by the Nuclear Regulatory Commission (NRC) through the requirements of Title 10 of the Code of Federal Regulations, Part 71. To meet the requirements of 10 CFR .sctn. 71, transportation casks must be designed to ensure criticality safety. The safety analyses for these transportation casks are presently based on the assumption that the fuel assemblies are unirradiated, i.e., the fissile content is the same as for the as-manufactured assembly. This assumption is conservative in the spent nuclear fuel case, as the fissile isotopes have been burned up as a result of the use of the fuel assembly in a reactor and, therefore, the fissile isotope content of the assembly is much lower than the as-manufactured content.
The capacity of transportation casks can be severely limited by the "fresh fuel" assumption, as larger criticality safety margins exist in the spent fuel assembly case. If credit could be taken for the burnup of the assemblies, cost savings in the transportation of spent fuel assemblies would result. In the development of canisters in support of dry storage of spent nuclear fuel, approval is currently being sought for a burnup credit methodology in support of package loading. The burnup credit methodology will rely on a combination of calculated burnup using reactor records, and burnup verification measurements to verify reactor records.
With increasing emphasis on issues related to the shipment of fuel for eventual disposition, burnup verification measurements and methodology are assuming a role of greater importance. The "fresh fuel" assumption results in very conservative designs for spent fuel racks, shipping canisters and waste repository storage. These overly conservative designs result in increased costs for the storage and shipping of spent nuclear fuel.
In order to take advantage of burnup credit for spent nuclear fuel, a method must be in place to reliably verify the fissionable content of fuel assemblies to ensure that criticality safety limits are not exceeded. Typical burnup verification methods require measurements of fuel assemblies to confirm reactor records of initial enrichment, burnup and decay time. These measurements rely on determining the neutron flux and, in some cases, the gamma dose in the vicinity of the fuel centerline.
The measurement technology in presently available commercial systems relies on .sup.235 U fission chambers to measure the neutron specific activity and either gamma ionization chambers, gamma scintillation detectors or solid-state semiconductor detectors (high-purity germanium HPGe), to detect gamma rays. Fission chambers and gamma ionization chambers are rather large gas-filled detectors. NaI(Tl) scintillation gamma ray detectors are typically large and require a photomultiplier tube and gamma ray shielding for operation in a spent fuel environment. HPGe gamma ray detectors require a liquid nitrogen cryogenic system or an electronic cooling system, since they are not capable of operation as high-resolution gamma ray detectors at higher temperatures. These detectors are sensitive to environmental factors such as temperature and the intense, mixed gamma ray and neutron field. For example, fission chambers are sensitive to gamma ray background, and gamma ionization chambers, NaI(Tl) detectors and HPGe detectors are all sensitive to neutron-induced background.
During the course of the use of a fuel assembly in a reactor core, higher actinides are produced by a chain of neutron captures followed by beta decay. In uranium fuel, the higher-actinide buildup chain originates with the .sup.238 U present in the fuel. Many of the higher actinides decay by spontaneous fission, a process which is accompanied by the emission of neutrons associated with fission. A secondary source of neutrons exists in oxide fuels where neutrons can be produced via the action of energetic alpha particles (primarily from the decay of higher actinides) on the .sup.18 O isotope of oxygen. It has been demonstrated by many workers that the neutron specific activity of spent fuel is related to burnup. Detailed mathematical relationships between the neutron emission rate and burnup have also been inferred using measurements on spent fuel assemblies. The functional form of this relationship is that the neutron emission rate is a function of assembly burnup raised to a power. Variables that affect the neutron emission rate include fuel type, initial enrichment, power history and decay time since discharge of the fuel assembly from the reactor.
Although spontaneously fissioning plutonium isotopes and plutonium alpha emitters are the dominant source of neutrons during the first fuel operation cycle, longer reactor core exposure times result in the production of curium isotopes which become the predominant source of the neutron specific activity for the spent fuel assembly. Most of the neutron emission will result from .sup.242 Cm (163 day half life) and .sup.244 Cm (17.9 year half life). For decay times more than a few years, .sup.244 Cm will be the major source of the neutron specific activity of a spent fuel assembly.
For shorter decay times, the neutron activity of .sup.242 Cm must be taken into account. Although the functional form generally covers all fuel assemblies of a particular design with different exponents for different design types, the curve will shift with initial enrichment. Therefore, knowledge of both the initial enrichment and time since discharge (decay time) are needed to accurately relate the observed neutron emission rate to burnup.
Typically, a combination of neutron measurements and reactor records are used to determine fuel burnup. In some cases, gamma ray measurements of fission product isotope gamma rays (primarily .sup.137 Cs) are used as a check on decay time. Either the gross gamma ray decay rate divided by the neutron emission rate can be related to groups of assemblies with common discharge times, or the .sup.134 Cs to .sup.137 Cs gamma decay rate ratio is measured directly to determine decay time. .sup.134 Cs has a half life of 2.06 years, and .sup.137 Cs has a half life of 30.1 years, so the decay rate ratio will change rapidly with time over a zero- to 20-year time period after discharge of the fuel assembly from the reactor. A measurement of either the .sup.134 Cs/.sup.137 Cs gamma emission ratio or the gross gamma emission rate is needed to verify the time since discharge (decay time) for the assembly. In the .sup.134 Cs/.sup.137 Cs case, the gamma ray intensity ratio provides a direct measure of the decay time. In the gross-gamma case, the assumption is made that most of the observed activity is .sup.137 Cs, and the gross gamma to neutron ratios allow the assemblies to be separated into groups according to common discharge times. The exact decay time is then determined from fuel assembly records.
Whether in support of pool storage or loading for dry storage, the measurements are conventionally carried out under water on isolated fuel assemblies which are raised from the fuel storage rack with an overhead crane. The detector fixture is designed to attach reproducibly to the fuel assembly, and normally measurements are carried out at the fuel center line with simultaneous measurements taken on opposite fuel flats to correct for asymmetries in the neutron emission rates.
The burnup verification methodology generally requires measurements to be performed on a set of fuel assemblies of a given type. The functional form for the neutron response as a function of burnup is established on the basis of at least three measurements and is updated as data from newly measured assemblies are added. Outliers are identified on the basis of agreement with the predictions of the fit function (usually, greater than three standard deviations from the predicted value is grounds for rejection), and identified for further study. Either incorrect records or a problem with the measurement could be responsible for outlier data points.
A common feature of all of the systems presently in use is that dependence is placed on a single measurement at one axial location. The neutron emission rate at that location depends on the average axial power profile for the reactor. While some of the presently available systems are capable of measurements at multiple axial locations, a sequence of adjustments of the relative position of the detectors and assembly is required for each measurement at each axial location. This measurement process entails measurement times roughly proportional to the desired number of axial positions and greater risk of fuel damage due to the large number of movements involved.
The present invention has been developed in view of the foregoing and other deficiencies of the prior art.